As prior art neutron individual dose meters using semiconductor detection elements, there are such that have been described in Radiation Protection Dosimetry, Vol. 27, No. 3, pp145-156 (1989) (hereinafter referred to as the prior art 1) and in U.S. Pat. No. 322787 as the prior art 2). Because the (hereinafter referred to semiconductor detection elements cannot detect neutrons directly, they manage to detect indirectly the neutrons by means of detecting charged particles which are generated by the interaction of the neutrons with other substances. For this purpose, the structure of the neutron detector in the prior art referred to in the former has a boron layer on the surface of the semiconductor detection element for detecting low energy thermal neutrons, and a polyethylene layer formed on the boron layer for detecting high energy fast neutrons. Further, in front of them, neutron moderator material is disposed to moderate neutron energy. The neutron detector in the prior art referred to in the latter case has a boron layer formed on the surface of a semiconductor detection element likewise the former example, the periphery of which layer is further surrounded by neutron moderator material in order to detect fast neutrons.
On the other hand, many of the prior art neutron exposure dose rate meters for monitoring neutrons in radiation handling facilities employ BF.sub.3 counters or .sup.3 He counters as described in the Japanese Patent Publication 63-235646 (1988) (hereinafter referred to as the prior art 3). It is disclosed, further, in the same publication that, in order to cut thermal neutrons and detect only fast neutrons, said counter is surrounded by neutron moderator, the surface of which is further surrounded by thermal neutron absorber, still further the surface of which is covered by neutron moderator, respectively.
Each country concerned is required to provide for various radiation detectors which satisfy energy response characteristics so that the effective dose equivalent evaluation conforming to the recommendation of the International Committee on Radiological Protection (ICRP) may be performed. Also in Japan, in accordance with the ICRP recommendation, the domestic radiation hazard prevention laws and regulations have been revised in April, 1989. Generally, radiation damages (dose quantities) differ depending on materials, even when they are exposed to radiation of the same energy. The so-called effective dose equivalent refers to a dose value for evaluation which precisely reflects an exposure quantity to neutrons of a human body. In order to implement this effective dose equivalent, it is necessary to survey and evaluate respective quantities of dose of each energy in a human body over a wider spectrum range of energy existing in radiation handling facilities. The range of energy includes a region for thermal neutrons having energy in the thermal neutron region below 0.5 eV, to fast neutrons ranging above 0.5 eV up to 10 MeV. Here, a sensitivity curve with respect to each specific energy is called a response. Since the difference in the effective dose equivalent responses in terms of sensitivities between the thermal neutron region and the MeV region is as large as more than 50 times, it is extremely difficult to implement this required response. The required response will be called as a dose equivalent response hereinunder. In order to satisfy the dose equivalent response, it is important (1) to make the shapes of sensitivity curves to coincide with each other or minimize the difference in the sensitivities, and (2) to increase sensitivity to each energy.
First, the above case (1) will be considered from a viewpoint of implementing a neutron individual dose meter.
In the prior art 1, in order to render the sensitivity curves to coincide with each other, in addition to the above-mentioned neutron detector, another neutron detector which has only a polyethylene layer on a semiconductor detector is provided therewith, then the responses of the two neutron detectors are added. However, as shown in FIG. 11 in page 155, the response performance has not been satisfied in an energy range from 10 K eV up to 1 MeV. In addition, there is a problem that the device tends to become large-sized because it utilizes two neutron detectors, and its processing circuitry is more complicated.
Next, we will consider the case (2). The prior arts 1 and 2 have a structure such that a layer of boron 10(.sup.10 B) is formed on the surface of a semiconductor detection element, and charged particles (.alpha. rays) generated in the layer when neutrons enter therein are detected by the semiconductor detection element. Such materials which generate charged particles when thermal neutrons enter will be called a converter hereinafter. In the neutron detector according to the prior art 1, the thickness of a boron 10 layer is formed to 1 micron m thickness by plasma doping techniques. Boron, however, has drawbacks that its melting point is as high as 2300 degree C., and its processibility is extremely poor. Further, the thickness of a boron film cannot be increased more than 1 micron m in order to prevent peel-off of the film due to temperature changes, because thermal expansion coefficients between boron and silicon which constitutes a semiconductor detection element differ as large as 3.5 times. The number of .alpha. rays generated in boron being increased in proportion to its film thickness, there has been a problem in implementing a neutron detector which has a sufficient detection sensitivity. While in the prior art 1 there has been described diffusion injection means for injecting boron, it has not been successful in increasing boron concentrations in diffused layers, thus failing to attain a neutron detector having a sufficient detection sensitivity. Boron enters into a nuclear reaction with a high probability when it encounters thermal neutrons, generating a lot of .alpha. rays, but it does not respond so vigorously to fast neutrons (high-energy neutrons over several ev). Therefore, it has a structure such that a polyethylene layer is provided, as has been described in the prior art 1, so as to enable detection of charged particles (protons) to be generated when fast neutrons enter therein by the semiconductor detection element. Such substance which generates protons when encountered with incident fast neutrons is called a proton radiator. It, however, has had a problem that it could not improve sensitivity to neutrons in an energy range from 10 KeV to 1 MeV.
Now, let's consider the neutron dose rate meter. According to the prior art 3, the dose equivalent response has not been satisfied. Further, because of a higher operating voltage at the time of neutron detection, which results in a more complicated measuring circuits, the prior art 3 has had a problem that it somewhat lacked reliability.
Further, the prior arts 1 and 2 have not been contemplated to be applied to neutron dose rate meters, hence there has been devoid of any such consideration.
A first object of the present invention is to provide for a portable neutron individual dose meter which attains the dose equivalent response.
A second object of the present invention is to provide for a portable neutron individual dose meter which attains the dose equivalent response by using a single semiconductor detection element.
A third object of the present invention is to provide for a portable neutron individual dose meter which has a high sensitivity to thermal neutrons.
A fourth object of the present invention is to provide for a portable neutron individual dose meter which has a high sensitivity to fast neutrons.
A fifth object of the present invention is to provide for a neutron exposure dose rate meter which implements the dose equivalent response.